The present invention refers to a method and a device for evaluating the integrity of the nuclear fuel in a nuclear plant having a reactor enclosing a reactor core comprising a number of fuel cladding members in which said fuel is present.
A core in a nuclear reactor comprises a plurality of fuel assemblies which are vertically provided with a certain distance from each other. The core is enclosed in a reactor vessel and submerged in a liquid, i.e. reactor water functioning as cooling medium as well as neutron moderator and flowing through the reactor vessel. The nuclear reactor also comprises a plurality of control rods which by being displaced into and out of the core regulate the effect produced by the reactor, and shut down and start, respectively, the reactor. A fuel assembly comprises a plurality of fuel rods vertically provided and each containing a staple of circular cylindrical, so called pellets of a nuclear fuel. Each fuel rod comprises a cladding tube which normally is manufactured in a zirconium alloy and which encloses the staple of fuel pellets. During an ongoing nuclear reaction, i.e. during the burn-up of the nuclear fuel, fission gases are formed which comprise radioactive inert gases and which normally are maintained within the fuel rod by means of the cladding tube. Such fission gases comprise different nuclides of xenon and krypton.
During operation it sometimes happens that a defect appears on a cladding tube. The first defect which appears on a cladding tube, for instance as a consequence of mechanical wear, is called a primary defect. The primary defect consists of a relatively small hole or a relatively small crack through which fission gases may be released. Such a primary defect is normally not particularly serious as such. However, after a period of time a primary defect may develop to a larger defect, a so-called secondary defect, which is more serious and consists of a larger hole or a crack extending axially along or essentially perpendicular to the longitudinal direction of the cladding tube. Such a secondary defect may arise by the penetration of water and steam into the fuel rod at the primary defect and by water and steam penetrating into the fuel rod at the primary defect and, by hydration, making the cladding tube brittle which after a certain period of time cracks during thermal or mechanical load or due to tensions caused by the hydrides themselves. By a secondary defect, also other fission products, such as iodine and caesium, are, except the fission gases previously mentioned, released. In case of a really serious fuel defect, for instance due to the fact that a fuel cladding tube is broken, also uranium and plutonium may start to leak out into the core. In order to prevent the occurrence of a more serious defect, it is of course important to detect a fuel defect at an early stage and adapt or interrupt the continuing operation of the reactor.
The defects on the cladding tubes may have many different grounds. In case of a too fast effect increase in the fuel, a combination of tension, due to the thermal extension, and chemical influence on the inner side of the cladding tube may result in the cracking of the cladding tube. This type of defect is called PCI (Pellet Cladding Interaction).
Another primary defect type is a so-called, debris defect, i.e. debris such as metal chips or the like, which causes wear defects on the cladding tubes from outside. Experience has shown that for instance in connection with repair and service of a nuclear reactor, such debris may be introduced and thereafter be carried by the water circulating through the core and give rise to a wear defect in the form of a primary defect on the cladding tube. A further type of defect may depend on a manufacturing failure. Finally, a so-called Dry-Out is also to be mentioned, which means that the liquid film which normally always is present on the outer side of the cladding tubes is evaporated. This leads to a quick local temperature increase in the fuel rod. Such a temperature increase leads to the melting of the cladding tube if the operation of the reactor is not immediately interrupted.
The development from a primary defect which is not serious as such to a secondary defect is influenced and may be accelerated by the manner in which the reactor is operated. If, for instance, the reactor is operated with a high effect or if many control rod movements are performed in the proximity of a defect, the defect may develop faster than it would have done at a lower reactor effect or if the control rods would have been maintained in a position. By stopping the reactor after a primary defect has occurred, the water flow into the defect fuel rod is increased and the degradation, i.e. the development towards a secondary defect, is accelerated.
If a fuel defect occurs and is detected during operation of a nuclear reactor, the people responsible for the operation may choose between shutting down immediately the reactor and replacing the defect fuel rod, or continuing the operation of the reactor, possibly by a locally reduced effect, until the next point of time for service and fuel replacement when the reactor is to be shut down. Normally, i.e. if no serious fuel defects have been detected, such a fuel replacement takes place once a year. The first alternative, i.e. to shut down the reactor, is very costly and ought to be avoided except at a serious fuel defect when the core may be more contaminated than is tolerated by the limits defined. In case of a PCI-defect, it would be sufficient to reduce the effect in the part of the core where the defect fuel rod is located. If the fuel defect is a debris defect, the reactor may be further operated at full effect until the next fuel replacement.
In order to determine if the reactor is to be shut down, if an effect reduction is sufficient or if it is possible to continue the operation at full effect, it is necessary to know the actual type of defect, where the defect is positioned, the size of the defect and the risk of degradation of the defect. These questions are very difficult to answer in a secure and reliable manner by the technique available today. Another problem in this connection is the determination if the primary defect has developed to a secondary defect or if a further primary defect has occurred in the case that higher activity levels of the reactor off-gases have been measured. Today, it is for the people responsible for the chemistry and the operation, in co-operation, to evaluate the type of the defect and to take a decision whether and how the reactor is to be operated further based on the activity levels measured and their experiences.
One way of detecting a fuel defect is to measure the total percentage of the radioactive inert gases in the reactor off-gases. One problem in this connection is that there are two different sources of radioactive inert gases, partly from the fuel due to a defect on the cladding tube and partly due to core contamination, i.e. from fissionable compounds being deposited onto different surfaces in the reactor core. By an increased number of fuel defects, which have occurred previously, this contamination increases and is therefore a growing source of error. However, the inert gases released from the core contamination contain a higher percentage of short-lived nuclides than the gases from a fuel defect.
Devices for detecting the presence of a fuel defect are known, for instance such a device is disclosed in U.S.Pat. No. 5 537 450. In this document, a device is disclosed which detects fuel defects on-line, i.e. during reactor operation, by leading a part of the off-gases from the reactor via a gammaspectrograph which continuously measures the nuclide composition and the activity level of the off-gases. It is also known to determine the position of a fuel defect by a method called flux-tilting, which involves the regulation of one control rod at a time in such a manner that the effect is changed locally in the core at the same time as the activity level in the off-gases is measured. An increase of the activity level in the off-gases may be noted during control rod movements in the proximity of the fuel defect. In such a manner, the fuel defect may be located. However, during the time of location, the effect of the reactor has to be reduced to between 60 and 80% of full effect. The reduction of the effect of the reactor to such a degree is costly and is to be avoided, if possible. Therefore, it is of course essential to realize if it is possible to postpone such a flux-tilting operation to a point of time when the reactor effect is to be reduced anyway. Furthermore, this method has the disadvantage that the control rod movements as such may deteriorate a defect already existing or accelerate the development of the defect towards a secondary defect.
The object of the present invention is to provide a method enabling a minimizing of the consequences of a nuclear fuel defect in a nuclear plant. Furthermore, an object of the present invention is to determine an essentially optimal way of continuing the operation of the plant in essentially each operation situation. According to a further aspect of the invention, an object is an automatic method, i.e. an automatic determination of the way of continuing the operation, which thus is not related to the individual.
This object is obtained by the method initially defined and comprising the steps of:
determining the value of each of a number of first parameters including activity data of the off-gases from the reactor,
determining the value of each of a number of second parameters including chemical data and activity data of the liquid flowing through the reactor,
determining the value of each of a number of operational parameters including at least one of the reactor effect, control rod position, actual fuel type, flow of off-gases, and flow of the liquid flowing through the reactor,
standardizing said values of the first and second parameters in relation to said values of the operational parameters,
analysing said standardized vales and said values of the operational parameters, and
estimating the integrity of the fuel based on said analysis.
By such a method which in an automatic manner takes into consideration a very large amount of available data of the function of the reactor, it is possible to obtain a relatively reliable knowledge of the integrity of the nuclear fuel during operation of the reactor. The integrity of the nuclear fuel means how the fuel and its possible substances and by-products, which are produced as a consequence of the ongoing nuclear reaction, are retained and will be retained in the fuel cladding members intended therefore. Thus, the integrity of the nuclear fuel is influenced by defects and crack formations of the fuel cladding members, the ageing of the fuel cladding member, as well as any other fuel leakage. From such a knowledge, it is then possible to determine if and how the reactor is to be operated further and according to an advantage embodiment of the invention, the method comprises the step of producing guidelines for the continuing operation of the plant with regard to said estimation. Consequently, by means of this knowledge it is possible to determine that a fuel defect has occurred and if the defect is of such a kind that, for instance, a flux-tilting operation is to be employed in order to exactly locate the defect for initiating a local effect reduction or enable replacement of the defect fuel cladding member.
According to a further embodiment of the invention, the method also comprises controlling the plant with regard to said guidelines. Thus, in accordance with the invention it is possible to control at least some functions in response to said guidelines in an automatic manner. Thereby, it is advantageous that said guidelines are produced essentially continuously at least during the operation of the plant. In such a manner, the guidelines may always correspond to the most actual values of said parameters. If, in addition, according to a further embodiment of the invention, the determining of the value of said first parameters is performed essentially continuously during the operation of the plant, the actual condition in the reactor with respect to the percentage of fission gases in the off-gases of the reactor may always form the basis for the continuing operation of the reactor.
According to a further embodiment of the invention, said analysis is performed with regard to historical data concerning said parameters. Such historical data may advantageously derive from the actual reactor but may also comprise data deriving from experiences gained during operation of other reactors. Advantageously, such historical data are obtained from a memory unit in which they are stored. Furthermore, said values may be supplied to the memory unit for updating and completion thereof. Such a memory unit, the data bank of which in such a manner may be expanded to be more and more extensive the longer the reactor is operated, consequently functions as a kind of reference library by which an appearing defect condition may be compared with previous similar conditions in the reactor. Advantageously, said values are supplied to the memory unit essentially continuously during the operation of the plant.
According to a further embodiment of the invention, said analysis comprises a comparison of said standardized values and said values of said operational parameters with said historical data and predetermined data. Such predetermined data may for instance include different limit values which may not be exceeded or which involve a determined measure if exceeded. By such a comparison, it is thus possible to take into consideration all available data for determining a fuel defect and the type of defect. However, it is to be noted that in order to perform the method according to the invention it is necessary to give an exact answer to the actual type of defect. It is rather essential to understand the extension and the expected development of the defect, which may form the basis for the continuing operation of the reactor.
The above-defined object is also obtained by the device initially defined and comprising:
first means arranged to enable the determining of the value of each of a number of first parameters including activity data of the off-gases from the reactor,
second means arranged to enable the determining of the value of each of a number of second parameters including chemical data and activity data of the liquid flowing through the reactor,
third means arranged to enable the determining of the value of each of a number of operational parameters including at least one of the reactor effect, control rod position, actual fuel type, flow of off-gases, and flow of the liquid flowing through the reactor,
fourth means arranged to enable the standardizing of said values of the first and second parameters in relation to said values of the operational parameters,
fifth means arranged to enable the analysing of said standardized values and said values of the operational parameters, and
sixth means arranged to enable the estimating of the integrity of the fuel based on said analysis. By such a device, it is possible to determine a fuel defect in a very reliable manner and by an at least reduced need of a subjective, individual-related judgement.